In presently known nuclear reactors, particularly in boiling water reactors, the reactor core typically contains a plurality of vertically oriented fuel assemblies arranged in an array such that a self-sustaining nuclear fission reaction can be controlled. The typical core is generally contained in a pressure vessel and submerged in a liquid such as water. The water may serve as both a core coolant and a neutron moderator. A series of vertically oriented, moveable control rods, composed of neutron absorbing material, are insertable between the fuel assemblies such that core reactivity control may be accomplished by adjusting the control rods within the core. In the usual case, water flows through channels located amongst the fuel rods and control rods in the core and is pumped upwardly from a lower plenum below the core to an upper plenum above the core.
To monitor the power distribution within the core, it is common practice to place in-core neutron detectors both radially and axially throughout the core. The signals from these neutron detectors are used to monitor core conditions and to initiate action in the event of a detected abnormality in the reactor. Core monitoring may include providing a reactor operator with analog or digital indications of the monitored conditions and providing an alarm when an abnormal condition is detected. Action, such as shutdown of the reactor, may also be automatically initiated when an abnormal core condition is detected.
A boiling water reactor is capable of entering a state called thermal-hydraulic instability that can challenge established fuel thermal and mechanical safety limits. Examination of the thermal-hydraulic stability of a reactor must be performed to prevent potential damage to the core. Thermal-hydraulic instability may be described as follows. Pressure perturbations at the core inlet cause flow disturbances that travel up the fuel channels as time-varying density waves. These waves result in local deviations from the steady-state axial pressure drop distribution. The local pressure drop in a fuel assembly is highly dependent on void fraction. Since the coolant voiding increases axially with greater core elevation, the highest void fraction is found at the channel outlet.
The effect of density waves on the total channel pressure drop is therefore effectively delayed in time--the void sweeping time--until the perturbation is felt at the channel exit. When the channel pressure drop time delay (phase lag) nears 180 degrees out of phase with the channel inlet flow variations, the fuel assembly can become thermal-hydraulically unstable. Thus, the thermal-hydraulic stability margin of a fuel channel is dependent on the phase lag caused by void sweeping time, and the gain which is dependent on the channel void distribution.
An additional complexity is introduced in boiling water reactor stability because of the reactor power dependency on coolant density. Local void reactivity responds to the time-varying density wave described above. The reactivity change affects local neutron flux and is manifested after a time delay as changes in fuel cladding surface heat flux and ultimately in local coolant voiding. This mechanism can also provide positive feedback to density wave oscillations. The neutronic feedback gain is dependent on how closely the fuel thermal time constant approximates the void sweeping time, and on the local void fraction.
The two feedback mechanisms, thermal hydraulic and neutronic, are coupled in a boiling water reactor core and produce oscillations in both core flow and thermal power. These oscillations can affect margins to fuel safety limits. In addition, core instabilities can occur even when neither feedback mechanism alone is sufficient to generate power oscillations.
Generally, existing thermal-hydraulic instability detection systems do not have the means to rapidly and accurately notify an operator of the core's stability margin. In the current invention, a simulated decay ratio signal is generated that relates to the thermal-hydraulic stability of a nuclear reactor which is input into means to take corrective action, notify an operator, or initiate an automatic suppression function upon reaching a minimum specified stability margin. A computer-based system is used, utilizing algorithms that provide a fast system response time such that a reliable indication of reductions in thermal-hydraulic stability of a core may be obtained prior to the core actually becoming unstable, and so that reactor operators may have sufficient time to take compensatory measures, or an automatic suppression function may be initiated.